New development of Subchannel CFD for complex flows in nuclear reactor system

This thesis presents new developments of a coarse-grid CFD method known as Subchannel CFD (SubChCFD) for the thermal-hydraulic analysis of nuclear reactors. The focus is on extending the capability of SubChCFD in capturing complex phenomena such as variable-property and buoyancy effects under supercritical conditions, flow swirl phenomena in wire-wrapped rod bundle configuration, and flow instabilities in the natural circulation loop (NCL). While traditional system and sub-channel codes offer computational efficiency, they are not designed to resolve local flow features. In contrast, high-fidelity CFD methods provide detailed insights but are computationally intensive, particularly for large-scale configurations. For supercritical conditions, developments include the implementation of empirical heat transfer and friction factor correlations for mixed convection and entrance region effects. The developments are validated through comparisons with resolved CFD simulations and available experimental data, showing reasonable accuracy across a range of operating conditions. Further developments are made to enable simulation of wire-wrapped fuel assemblies by incorporating momentum source terms to account for the influence of helical spacers. Furthermore, the SubChCFD is applied to natural circulation loop systems to assess its ability to reproduce flow instabilities. In summary, the work reported herein has extended the capability of SubChCFD to deal with a number of complex flows encountered in nuclear reactors successfully, and has demonstrated that SubChCFD captures key thermal-hydraulic features, such as heat transfer enhancement and deterioration, secondary flow structures, and oscillatory flow behaviour. This work further establishes SubChCFD as a practical alternative for a nuclear thermal-hydraulic analysis tool.